ML20127E053

Requests Addl Info Re June 1992 Application for Design Certification of AP600 to Complete Review

ML20127E053
Person / Time
Site:	05200003
Issue date:	11/16/1992
From:	Kenyon T Office of Nuclear Reactor Regulation
To:	Liparulo N WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP.
References
NUDOCS 9301190141
Download: ML20127E053 (18)

Text

• -

• hAbd h kC, g,* $8cg y

ff o g UNITED STATES NUCLEAR REGULAlORY COMMISSION 5

/(,9 gE WASHINGTON, D. C 20555 November 16, 1992

% . . . . . *[

4 Docket No.52-003 Mr. Nicholas J. Liparulo Nuclear Safety and Regulatory Activities Westinghouse Electric Corporation P.O. Box 355 Pittsburgh, Pennsylvania 15230

SUBJECT:

REQUEST FOR ADDITIONAL INFORMATION ON THE AP600 As a result of its review of the June 1992 application for design certifica-tion of the AP600, the staff has determined that it needs additional informa-tion in order to complete its review. The additional information is needed in the areas of testing (100.7)," auxiliary systems (Q410.16-Q410.25), flood protection (Q410.26-Q410.50), missile protection (Q410.51-Q410.75)', piping failures (outside containment) (Q410.76-Q410.92), instrumentation and controls (Q420.8), and radioactive waste management (Q460.8-Q460.16). Enclosed are the staff's questions. Please respond to this request within 120 days of the date-of receipt of this letter.

You have requested that portions of the information submitted in the June 1992 application for design certification be exempt from mandatory public disclo-sure. While the staff has not completed its review of your request in accordance with the requirements of 10 CFR 2.790, that portion of the submitted information is being withheld from public disclosure pending the staff's final determination. The staff concludes that this request for additional information does not contain those portions of the information for which exemption is sought. - However, the staff will withhold this letter from public disclosure for 30 calendar days from the date of this letter to allow Westinghouse the opportunity to verify the staff's conclusions. If, after that time, you do not request that all or portions of the information in De enclosures be withheld from public disclosure in accordance with 10 CFR 2.790,~

this letter will be placed-in the NRC's Public Document Room.

• The numbers in parentheses designate the tracking numbers assigned to g the questions. ,

t 9301190141_951156' 3 i% ~.

" i PDR ADOCK 0520

Mr. Nicholas J. Liparulo November 16, 1992 The reporting and/or recording requirements contained in this letter affect fewer than ten respondents; therefore, 011B clearance is not required under P.L.96-511.

If you have any questions regarding this matter, you can contact me at (301) 504-1120.

Sincerely, Origina! Signed By:

Thomas J. Kenyon, Project Manager Standardization Project Directorate Associate Director for Advanced Reactors and License Renewal Office of Nuclear Reactor Regulation

Enclosure:

As stated cc w/ enclosure:

See next page DISTRIBUTION:

• Central File PDST R/F TMurley/FMiraglia DCrutchfield

• PDR WTravers RPierson RBorchardt TKenyon RHasselberg GGrant, EDO JMoore, 15B18 ACRS (10) MSiemien, 15B18 Cli, 8D1 JRaval, 801 --

JLyons, 8D1 TChandrasekaran, 801 RArchitzel, 801 BBurton, 8D1 GHubbard, 801 HLi, SH7 MChiramal, 8H7 PShea OFC: LA:PDST:ADAR PM:PDST:ADAR PM:PDST,:ADAR SC:PDSJ;fADAY NAME: PShea gg TKenyf[g R$ RBorhrdt DATE: 11/[p/!i 11/g/92 11/4/92 11h92  ;

0FFICIAL RECORD COPY:

DOCUMENT NAME: LETTER.VI

• To be held for 30 days I

Mr. Nicholas J. Liparulo Westinghouse Electric Corporation Docket No.52-003 AP600 cc: Mr. B. A. McIntyre Advanced Plant Safety & Licensing Westinghouse Electric Corporation Energy Systems Business Unit Box 355 Pittsburgh, Pennsylvania 15230 Mr. M. D. Beaumont Nuclear and Advanced Technology Division Westinghouse Electric Corporation One Montrose Metro 11921 Rockville Pike Suite 350 Rockville, Maryland 20852 Mr. Daniel F. Giessing U. S. Department of Energy NE-42 Washington, D.C. 20585 Mr. S. M. Modro EG&G Idaho Inc.

Post Office Box 1625 Idaho Falls, Idaho 83415

ENCLOSURE REQUEST FOR ADDITIONAL INFORMATION ON THE WESTINGHOUSE AP600 DESIGN TESTING 100.7 Provide written responses to the pre-application requests for additional information that were requested in letters dated January 30, 1992, June 22, 1992, July 21, 1992, and September 1, 1992, or-provide a cross-reference to any responses that may have already been formally addressed.

AUXILIARY SYSTEMS Reactor Coolant Pressure Boundary Leakaae Detqction 410.16 Section 5.2.5.1.2 of the SSAR states that limits for reactor coolant leakage are identified in the technical specifications. Section 3.4.7 of Chapter 16 of the SSAR (Technical Specifications) specifies the limits of identified and unidentified leakage. Describe how the identified leakage can be quantitatively measured, and how operators can determine if the leakage limit is exceeded.

410.17 -Position C.9 of RG 1.45 states that the technical spec;fications should address the availability of various types of instruments for RCPB leakage to ensure adequate coverage at all times. Describe how the AP600 design will meet this regulatory position (Section 5.2.5).

410.18 Position C.8 of RG 1.45 states that the _ leakage detection systems should be equipped with provisions to readily permit testing-for operability and calibration during plant operation. Describe how the AP600 design will meet this regulatory position (Section 5.2.5).

410.19 Position C.7 of RG 1.45 states that procedures for converting various leakage indication to a common leakage equivalent should be available to the operators. Describe how the TP600 design will implement-this regulatory position (Section 5.2.5).

410.20 Position C.6 of RG 1.45 states that the leakage detection systems-should be capable of performing their functions following seismic events that do not require plant shutdown. The airborne particulate radioactivity monitoring system should remain functional when subjected to the SSE. Describe how the AP600 design will meet - this -

regulatory position (Section 5.2.5).

Pressurizer Relief Discharae System 410.21 Section 5.4.11 of the SSAR states that the safety valves connected to the top of the pressurizer provide for overpressure protection of the reactor coolant system. The discharge of the safety valves is connected through a rupture disk to containment atmosphere. The

l, .,

discharge is directed away from any safety related equipment, structures, or supports that could be damaged to the extent that emergency plant shutdown is prevented by such a discharge.

Identify the worse case steam / water discharge and provide an analysis to determine the distance that the discharge jet may cause impingement damage. Provide a diagram of the proximity of the safety valves-to show all the equipment, components, structures, or supports within this distance.

410.22 Section 5.4.11.2 of the SSAR states that the piping and instrumentation diagram (P&lD) for the connection between the automatic depressurization system valves and the in-containment refueling water storage tank is shown in Figure 6.3-1. Figure 6.3.1 is the P&l0 for the passive core cooling system. It does not appear to contain the above mentioned information. Provide the appropriate reference for the pressurizer relief discharge system.

410.23 Section 5.4.11.2 of the SSAR states that the discharge of water, steam, and gases from the first-stage automatic depressurization system valves when used to vent noncondensable gases does not result in pressure in excess of the in-containment refueling water storage tank (IRWST) design pressure.

When the high pressure discharge of steam and noncondensable gases, through piping system, injects into the in-containment water storage tank, it will be condensed and mixed with the tank water. The IRWST, in this case, will function similarly to the suppression pool in the BWR plants. Provide an analysis to demonstrate that the hydrodynamic loads on the tank and piping have been adequately considered.

410.24 Section 5.4.11.4 of the SSAR, entitled " Instrumentation Requirements",

states that the instrumentation for the safety valve discharge pipe, containment, and in-containment refueling water storage tank are discussed in Sections 5.4,9, 6.2, and 6.3, respectively. It is not-clear where the specific information regarding the instrumentation requirements in the pressurizer relief discharge system is located in the referenced sections. Provide this specific information or provide a more specific reference for the information in the SSAR.

Control Room Habitability 410.25- Section 9.4.1 of the SSAR states that a supplemental air filtration subsystem filters outside makeup air and pressurizes the MCR and TSC-areas if high airborne radioactivity is detected in the MCR supply air duct and/or receipt of a containment isolation signal. Also, Note 2 in Figure 9.4.1-1 (sheet 1 of 6) of the 'SSAR states that the.

supplemental air filtration system may be deleted pending radiological analysis by Westinghouse and-is shown-for information only. Additionally, the referenced figure shows only one fresh air intake and it is not clear that any radiation monitors are provided.

_ ~

4 Is Westinghouse going to delete the supplemental air filtration system? If so, describe in detail the justifications for doing so.

To meet GDC 19, Westinghouse must demonstrate that-the air filtration system is a safety-related ESF filtration system, and that it conforms with the guidelines provided in Sections 6.4, 6.5.1, and 9.4.1 of the l SRP, Regulatory Guide 1.52, ASTM Standard 03803-1989, and ASME Standards N509-1989, N510-1989, and AG-1-1991, which includes-appropriate provisions for single-failure criterion design, dual fresh-air intakes, redundant radiation monitoring in each fresh air intake and associated testing requirements of the safety-grade components.

Also, provide an updated flow diagram and -general arrangement drawings demonstrating conformance with the guidance of the referenced SRP sections, regulatory guide, and standards.

FLOOD PROTECTION 410.26 Section 3.4.1.2.2,1 of the SSAR states that reverse flow from the containment sump to the two PxS compartments and the CVCS compartment is prevented by redundant " backflow preventers" in each of the three compartment drain lines. Provide design information on these components, including leakage characteristics. Discuss the likelihood of failure of these components and the subsequent flooding effect.

410.27 Identify all safety-related equipment and equipment important-to-safety (i.e. non-safety related equipment whose failure could adversely affect the ability of safety related equipment to perform its safety function) requiring protection from internal and external flooding (Section 3.4.1).

410.28 Identify potential sources of internal flooding on a floor-by-floor basis in all buildings containing safety-related equipment. How will safety related equipment and equipment important-to-safety be protected from flooding from these sources (Section 3.4.1)?

410.29 Identify potential sources of external flooding from components which are within the AP600 design scope (Section 3.4.1).

410.30 Describe if the maximum flood level in Table 2.0-1 of the SSAR takes into account probable _ maximum floods (PMFs) generated by a combination of probable maximum precipitation (PMP) or other combinations of less severe environmental and man-made events along with seismic and wind effects.

410.31 Throughout Section 3.4.1 of the SSAR, distinctions appear to be Nade between flood protection for safe-shutdown equipment versus safety-related equipment. Describe-how flood protection requirements differ between safe shutdown and safety-related equipment.

410.32 How will safety-related equipment and equipment important-to-safety be protected from failures of structures, systems, and components that are not within the AP600 design scope (Section 3.4.1)?

~ w -

e

l 410.33 Discuss the ability of safety-related equipment to perform its safety function while fully flooded, partially flooded, or wet (e.g. frcm spray). Particular attention should be given to the five containment isolation valves that are below the internal flood level, and are, therefore, subject to flooding (Section 3.4.1).

410.34 Section 3.4.1.2.2.1 of the SSAR states that the PXS-A, PXS-B, and CVCS compartments are physically separated and isolated from each other by structural walls such that flooding in any of these compartments or in the RCS compartment cannot cause flooding in any of the other compartments. This appears to contradict another statement in this section which says that, because the floor drains for these compartments are routed to the containment sump, flooding in one compartment could cause flooding in another compartment. The staff recognizes that " backflow preventers" are located in each line to prevent reverse flow into other compartments but insufficient detail has been provided on the design and operations of these components (see Q410.26). Clarify these statements.

410.35 Identify safety-related equipment and equipment important-to-safety that are subject to groundwater seepage, and discuss how this will be controlled (Section 3.4.1).

410.36 Identify whether ficod protection depends upon the use of a dewatering and drainage system. If so, provide details on the system (Section 3.4.1).

410.37 Section 3.4.1,2.2.1 of the SSAR states that the safe shutdown components located in PXS-A and PXS-B are redundant and " essentially identical." Clarify what is meant by " essentially identical."

410.38 How are txternal penetrations that are below plant grade protected from external flooding (Section 3.4.1)?

410.39 Discuss design criteria for doors, walls, and penetrations used to provide internal and external flood protection (Section 3.4.1).

410.40 Are any external or internal doorways or passageways too large to close with a single door? If so, how will leakage be prevented (Section 3.4.1)?

410.41 Provide design information en waterproofing membranes, waterstops, watertight doors, and other protective features (Section 3.4.1).

410.42 Two lines are routed from the IRWST to each of the PXS compartments.

The six inch line is routed to PXS-A and the 10-inch line is routed to PXS-B. What is the purpose of these lines? Why are these lines sized differently? What is the effect if the PXS compartment overflows (Section 3.4.1)?

410.43 Identify the component cooling water on the building layout drawings (Section 3.4.1).

l d

l 5-410.44 Discuss possible flood hazards resulting from below grade tunnels between buildings (Section 3.4.1),

410.45 Are monitors which detect flooding in areas containing safety-related equipment and equipment important-to-safety related (!,ection 3.4.1)?

410.46 Do any open-cycle systems enter any buildings housing safety-related .1 equipment and equipment important-to-safety? If so, how will this equipment be protected from the effects of a break 'n that part of the open-cycle system within the building (Section 3.4..)?

410.47 Identify all watertight doors and hatches on-the general arrangement ,

drawings, not just 3-hour fire doors (Section 3.4.1).

410.48 It appears from Figure 1.2-5 of the SSAR that both divisions of the spent fuel ' pool cooling pumps and heat exchangers are susceptible to flooding. How will these components be protected?

410.49 Identify the potential flooding consequences if the IRWST or the PRHR heat exchangers were to fail. Identify the protective features used to protect safety-related equipment and equipment important-to-safety from the resulting flooding (Section 3.4.1).

410.50 How will the remote shutdown panel be protected from external and internal flooding (Section 3.4.1)?.

4 MISSILE PROTECTION Internally-Generated Missiles (Outside Containment) 410.51. Identify all safety-related equipment and equipment important-to-

, safety (i.e. non-safety related equipment whose failure could adversely affect the ability of safety-related equipment to perform its safety function) that require protection from internally-generated

, missiles (outside containment) (Section 3-5.1.1). .

410.52 How will safety-related equipment and equipment important-to-safety be protected from-turbine missiles (Section 3.5.1.1)?

410.53 How will safety-related equipment and equipment important-to-safety outside containment'be protected from credible secondary missiles-(e.g., concrete fragments) (Section 3.5.1.1)?..

410.54 Identify safety-related equipment and equipment important-to-safety.

that are subject to missiles from non-seismic Category I structures, systems, and components, and discuss how-this equipment will be protected from such missiles (Section 3.5.1.1).

410.55 Provide sample analyses to demonstrate that housings of rotating equipment can contain missiles generated by the rotating equipment-(Section 3.5.1.1).

J

)

l 410.56 Provide examples of equipment features used to prevent missile i generation in the "special situations" referred to in Section 3.5.1.1.2.1 of the SSAR. ]

410.57 Provide sample analyses-to demonstrate that credible missiles cannot j be generated from non-high-energy fluid systems. Are.high energy systems that meet the 2% and 1% rules defined in Section 3.6.1.1 of-the SSAR considered high-energy systems from the standpoint of missile generation (Section 3.5.1.1)?

410.58 Provide design information regarding the retaining ring and yoke used in the valve bonnets of pressure seal, bonnet-type valves (Section 3.5.1.1).

410.59 Provide sample analyses to demonstrate that the stored energy for nuts, bolts, nut and bolt, and nut and stud combinations is not enough to generate credible missiles (Section 3.5.1.1).

410.60 Provide analysis to demonstrate that the contents of one hydrogen bottle will not lead to an explosion. Show how hydrogen release is limited to one bottle should a supply line fail (Section 3.5.1.1).

410.61 Identify the sources of missiles that meet the criteria in Section 3.5.1.1.2.3 of the SSAR.

410.62 How will the remote shutdown panel be protected from missiles generated outside containment (Section 3.5.1.1)?

Internally-Generated Missiles (Inside Containment) 410.63 Identify all safety-related equipment and equipment important-to- -

safety which require protection from internally-generated missiles (inside containment) (Section 3.5.1.2).

410,64 Identify all sources of primary and credible secondary missiles (concrete fragments) that could adversely impact safety-related equipment and equipment important-to-safety inside containment (Section 3.5.1.2).

410.65 Provide an analysis to demonstrate that the reactor coolant pump casing can contain a missile generated by the pump (Section 3.5.1.2).

410.66 Provide additional detailed information on how components inside containment are prevented from. producing credible missiles (Section 3.5.1.2.1.1).

410.67 Identify the sources of missiles meeting the criteria in Section 3.5.1.2.1.3 of the SSAR.

i

Missiles Generated by Natural Phenomena 410.68 Identify all safety-related equipment and equipment important-to-safety that require protection from missiles generated by natural phenomena, including the design basis tornado (DBT) (Section 3.5.1.4).

410.69 Discuss how safety related structures, systems, and camponents (SSCs) that are important-to-safety will be protected from missiles generated by natural phenomena, including the DBT (Section 3.5.1.4).

410.70 Provide an estimate of the strike probability per ear for the plant (Section 3.5.1.4).

410.71 How will the remote shutdown panel be protected from missiles generated by natural phenomena (Section 3.5.1.4)?

Externally-Generated Missiles 410.72 Identify all equipment important-to-safety that require protection from externally-generated ;ssiles (Section 3.5.2).

410.73 How will the passive containment cooling system (including tank, valves, piping) be protected from external missiles and their effects (Section 3.5.2)?

410.74 How will the fuel storage pool and the fuel within the pool be protected from. external missiles and their- effects -(Section 3.5.2)?

410.75 How will the remote shutdown panel be protected from external missiles (Section 3.5.2)?

PIPING FAILURES OUTSIDE CONTAINMENT 410.76 Provide a pipe break effects analysis (Section 3.6.1), including:

a. postulated piping failures (includingLthose identified in Section 3.6.1.1.F of the SSAR),

b. pipe failure locations (circumferential and longitudinal break,-

leakage cracks, and through-wall cracks), -

c. piping 1that meets the leak-before-break (LBB) criteria, and

d. protective structures and other features used to mitigate the consequences of the piping failures.

410.77 Will containment penetration areas meet the break exclusion provisions of B.1.b of Branch Technical Position MEB 3-1 (Section 3.6.1)?

. - , - -. , , - . . . . . . . - ~ ,-

l l

410.78 Section 3.6.1.1.F of the SSAR identifies the initiating events for the pipe f ailure ef fects analysis. Clarify if leakage cracks in moderate energy pipes ate considered as initiating events.

410.79 Clarify what is meant in Section 3.6.1.1.J of the SSAR.

410.80 Describe why the turbine stop valves and feedwater control valves are credited in the single failure analysis to limit a break of the main steam or feedwater lines inside containment (Section 3.6.1).

410.81 Identify the locations of all pipe whip restraints (Section 3.6.1),

410.82 Identify all line restrictions in high-energy lines that impact on the -

calcult.tions of thrust and jet impingement forces, and provide a sample calculation to demonstrate how these restrictions were addressed in these calculations (Section 3.6.1).

410.G3 How is the remote shutdown panel protected from the effects of pipe failures (Section 3.6.1)?

410.84 Provide examples of equipment that are subject to environmental and flooding effects due to venting from an adjoining subcompartment (Section 3.6.1).

410.85 Do any subcompartments inside the containment contain high-energy lines between 3 - 4 inches? If so, are breaks from lines of this size considered in the subcompartment pressurization analysis (Section 3.6.1)?

410.86 Provide more ietailed information regarding pressurization loads for the IRWST anc the reactor vessel annulus. What is the basis for a 5 -

gpm leakage rate in the primary loop piping (Section 3.6.1)?

410.87 Which pipe (s) are postulated to fall to provide the internal reactor pressure vessel asymmetric pressurization loads (Section 3.6.1)?

410.88 Provide the dimensions for the reinforced concrete walls separating the control room from the MSIV compartmC;t. Include wall dimensions on the plant arrangement drawings (Section 3.6.1) .

410.89 Provide the results of an analysis of pipe breaks in the turbine building and their effect on the control room and remote shutdown panel (Section 3.6.1),

410.90 Provide additional detailcd information regarding the methods used to protect the reactor coolant loops from the effects of failures of the main steam and feedwater lines and vice versa (Section 3.6.1).

410.91 Describe how safety-related instrumentation is protected from pipe failures and their effects (Section 3.6.1).

l

.g 410.92 Identify systems important-to-safety that-require protection from pipe failures and their effects (Section 3.6.1).

INSTRUMENTATION AND CONTROLS 420.8 The instrumentation and control system of the AP600 uses j nilcroprocessor based distributed digital equipment to perform plant protection and safety monitoring functions. The software design quality is a major issue which must be adequately addressed by Westinghouse for the staff to make a safety determination. Because microprocessor and digital control technology is rapidly evolving it is important that the certified design description and the ITAAC do not " lock in" a design that would be obsolete at the time of construction. The staff's approach in this area is to certify a design. process and " lock in" the specific design acceptance criteria (DAC). The degree to which a particular aspect of design is " locked-in" (Tier 1 or Tier 2) will be described in the certification rule.

The ITAAC submittal for the Protection and Safety Monitoring System of the AP600 only addresses functional requirements and does not provide design details or-the design process commitment. 'The Certified Design Commitment for the microprocessor-based digital protection system or digital control system should address the software design process-commitment and describe a formal design implementation process with a phased inspections, tests, analyses, and acceptance criteria. The ITAAC will be inspected by the NRC to verify conformance with the requirements at several phases or stages during the design process.

At each phase of the ITAAC, implementation of ~ the design development must be verified to be in accordance with the certified design process. Upon completion of each phase of the ITAAC, the COL holder will certify to the NRC that the stage has been completed and the design and construction completed up through that stage is in-compliance with the certified design. The COL holder will also provide a description of the next phase of design development and ; -

associated testing, analysis and acceptance criteria in.enough detail that the NRC staff can determine whether or not the proposed design development and testing is consistent with the certified design process and next ITAAC. This phased process will continue until all E !TAAC stages for all the safety related software are completed.

In addition to the design process' commitment,.the ITAAC for the Protection and Safety Monitoring System should 'also address the following elements:

a common mode failure prevention

= human factors aspect of the design

+-

communications (data link)

- bypass ~ capability (operation & maintenance modes)

= setpoint methodology

• safety action seal-in provision

• physical separation of channels

• on-line testing and surveillance testing

• EMI/RFI protection a equipment qualification

• safety and control systems-interactions

= cross reference to SSAR drawings RADIOACTIVE WASTE MANAGEMENT 460.8 Provide the followim: information regarding the source terms for evaluating the expucted performance of radwaste management systems (Sections 11.1 and 11.2):

a. Table 11.2-6 states that the shim bleed rate is 737 gallons per day (gpd). The staff has determined that this rate is 288 gpd using the value of the reactor coolant letdown flow given in Table 11.1-7 of the SSAR and 657.5 gpd calculated from the yearly average of the CVCS letdown given in Table 11.2-1 of the SSAR.

The staff has verified that a shim bleed rate-of 288 gpd gives-the same reactor coolant activity (RCA) values for noble gss radionuclides given in Table 11.1-8 of the SSAR. Resolve the inconsistencies among the tables,

b. The staff has determined that a primary-to-secondary leak rate of 75 lb/ day (NUREG-0017, Rev.1 value) rather than the 100 lb/ day given in Table 11.1-7 of the SSAR results in the steam generator steam activity values given in Table 11.1-8 of the SSAR for noble-gases. Correct the inconsistency.

c. Tables 11.1-7 gnd 11.2-6 of the 3SSAR give the total steam flow rate as 8.4x10 lb/hr and 8.4x10 lb/hr, respectively. The staff believes that the first value is correct. Lorrect the inconsistency.

d. The staff concludes that the steam generator liquid activities and, consequently, the steam generator steam activities of halogens, Cesium (Cs), Rubidium (Rb), and other nuclides listed in Table 11.1-8 of the SSAR are not consistent with those- that would be calculated using the method given in Revision 1 of NUREG-0017.

Describe how the activities were calculated, and correct them, if appropriate.

460.9 Provide the following information regarding the liquid radwaste management system (Section 11.2):

a. Section 11.2.2.1.1 of the SSAR states that any of the four ion exchangers provided in series for processing all liquid radwaste strezms can be manually bypassed. In addition, the SSAR does not provide sufficient test data that supports additional-radioactivity removal from waste streams by a third and a fourth ion exchanger in series with the first two ion exchangers.

Further, the staff notes that the additional credit due to the ion

l l

exchangers of the chemical volume control system for the shim- l bleed stream is already built into the code since such credit is used for calculating the primary coolant concentrations of radionuclides. For the above reasons, the staff estimates that  !

the following DFs appear more appropriate than the ones used in '

Table 11.2-6 of the SSAR:

3 Halogens 10 l Cs, Rb 20  ;

3 Others 10 l These values assume that there are at least-two mixed beds in series.  ;

In light of the above discussion, provide justification for the DFs in Table 11.2-6 of the SSAR or correct the values, as appropriate. in resolving the staff's concern regarding the DFs, clarify whether at least two of these four beds are mixed, because only a combination of two mixed beds'in series can provide optimum removal capability for different categories of radionuclides,

b. Section 11.2.3 of the SSAR states that, except for the steam generator blowdown wastes, all other processed liquid radwastes will be discharged to the environment rather than recycled within.

the plant. However, Table 11.1-8 of the SSAR shows a tritium RCA of lyCi/gm, which indicates that a moderate amount of tritium will be recycled (see Table 2-6 of Revision 1 to NUREG-0017). Further, the staff is concerned that, with minimum processing of Cs and Rb nuclides in the waste streams (DF of 20) and maximum discharge of the processed liquid radwastes, Cs and Rb radionuclide releases via liquid effluents can pose a problem in terms of compliance-with the concentration limits of 10 CFR Part 20 and the offsite dose limits of Appendix I of 10 CFR Part 50, unless there is substantial dilution of the waste streams-prior to their discharge. In light of the above discussion, describe why the AP600 is designed for minimum recycling, which is at variance with industry practice and standards (see Subsection 4.1.5 of ANSI /ANS 55.6) c, Collection. times for various wastes given in Table 11.2-6 of the SSAR appear to be inconsistent with the values that will result from using the methodology of Revision 1 to NUREG-0017. The NUREG recommends using only 80 percent of one collector tank's full-capacity as the fill volume provided there are t c. tanks of equal capacity. It is not clear whether the volumes of all tanks given in-Table 11.2-2 of the SSAR represent 80 percent of their full capacity. Justify or resolve the inconsistency,

d. The process and discharge times for the various liquid waste streams given in Table 11.2-6 of the SSAR (1. day for all of them) appear to be inconsistent .with the methodology of Revision 1 to NUREG-0017. The calculated process time cannot be further

.. - . .. ~ . -. - . - . ..

increased by half of the calculated discharge time when the monitor tanks have smaller volumes than their corresponding holdup or collection tanks. Further, the methodology recommends using only 80 percent of the full capacity of one monitor tank, provided that there are two monitor tanks of equal capacity. Justify or resolve the inconsistency,

e. The input from leakage of the spent fuel pit liner, the reactor containment cooling system, and the reactor coolant pump seal, and  :

the sampling drains of the secondary coolant system are either not I included in Table 11.2-1 of the SSAR or are much lower in the i subject table than the inputs from these sources given in Section 3.2.1 of Revision 1 to NUREG-0017 and ANSI /ANS-55.6. The staff recognizes that the product of the expected activity level and the daily input given in the subject table for the applicable sources discussed above is greater than the corresponding NUREG products.

However, the inputs have a bearing on the sizing of the liquid radwaste system equipment and collection and processing times.

Additionally, the equipment drain tank may also be a source (NUREG-0017 gives a much higher combined total waste generation rate for equipment drains and clean wastes than the AP600 design).

Provide the missing information. Provide the reasons for the inconsistencies or correct the values.

f. Appendix 1A of the SSAR demonstrates that the liquid radwaste management system for the AP600 design meets Position C.1.2 of RG 1.143 with respect to the design features for applicable tanks ,

(i.e., tanks located outside the containment and carrying radioactive materials). However, Table 11.2-3 of the SSAR, which lists the applicable tanks with level and alarm features, does not include the condensate storage tank (CST). Clarify whether the -

CST has level indication and alarm features in accordance with-Position C.I.2.1 of RG 1.143. Also, clarify whether the AP600-design includes the specific design features-discussed in Positions C.I.2.2 through C.I.2.5 uf RG 1.143 for all the-applicable tanks,

g. Clarify whether the AP600 design has a single liquid waste discharge path as shown in Figure 11.2-1 of the SSAR. Also, clarify whether the AP600 design permits different categories of wastes to be discharge simultaneously at any time, provided such cumulative discharge is within the applicable regulatory limits.

If not, identify the design features that preclude such a simultaneous discharge.

460,10 Provide the following information regarding the gaseous radwaste management system (Section 11.3):

a. Tables 11.2-6 and 11.3-1 of the SSAR give different holdup times for Xenon and Krypton in the charcoal delay beds. Clarify the discrepancy between the tables.

b. Describe the basis for the RCS degassing days (17.4 and 1.0 for Xenon and Krypton, respectively) given in Table 11.2-6 of the SSAR. The waste gas system releases given in Table 11.3-3 of the SSAR do not appear to be correct. Confirm the acceptability of this information or correct it, as appropriate.

c. Discuss the provisions for monitoring the individual performance of the equipment within the charcoal delay bed system. Include a

?"

list of alarmed process parameters for the delay bed system.

i 460.11 Provide the following information regarding the solid radwaste management system (Section 11.4):

a. Since solidification and encapsulation are not the same, clarify whether either of the above two options may be used for processing spent resins in addition to a third option, namely dewatering the resins (note that encapsulation is not generally used for i processing spent resins). Also, clarify whether the AP600 solid radwaste management system design deviates from the EPRI Requirements Document for passive reactor designs. The Requirements Document recommends only dewatering for processing the spent resins.

b. Identify the specific design features provided in the system design to comply with GDCs 60, 63 and 64 as they relate to (1) control of release of radioactive materials to the environment from the plant areas where the solid radwastes are processed, and (2) monitoring radiation levels and leakage,

c. Clarify whether the description and discussion of acceptability of the portable grouting unit that may be used for processing spent -

filters is within the COL applicant's scope. If it is within the AP600 design scope, provide specific details of the unit.

d. The staff is concerned that the projected (Tables 11.4-4 and 11.4-5 of the SSAR) annual solid radwaste volumes to be disposed (1729 CF for the expected case and 3843 CF for the maximum case) are significantly lower than that actually shipped volume for operating PWRs (EPRI NP-5528, February 1988, Volume 2 - Plants Without Evaporators for the Years 1985 and 1986: 9550 CF). The staff recognizes that the projected volume agrees with the value proposed in the EPRI Requirements Document (1750 CF per year).

The EPRI-proposed value depends on following what EPRI regards as sound design and operating techniques outlined in the document (Paragraph 8.1.2.2 of Appendix B of Chapter 12) for reducing the shipment of processed solid waste volume. One of the operating techniques is to avoid solidification and instead use only dewatering for solidifying the wet solid wastes. As stated above, the AP600 design includes solidification as one of the options.

The staff is concerned that the storage volume allotted for processed solid wastes may be inadequate if it is to be based on

s the projected shipment volumes given in the SSAR tables.

Therefore, provide justification for the projected volumes given in the subject SSAR tables or revise the values as appropriate.

e. Clarify why the AP600 design does not include phase separator tanks, as recommended in-the EPRI Requirements Document for passive reactor designs.

f. Section 11.4.1.3 of the SSAR identifies the capability to store processed and packaged solid wastes at the site for at least six months to account for possible delay or disruption of offsite shipping of the wastes as one of the design objectives of the solid waste management system. However, there is no description of the on-site storage facility in the SSAR. Provide a description of the facility, and clarify whether it conforms with the recommendations identified for such a facility in Section 5.4 of Chapter 12 of the EPRI Requirements Document for passive reactor designs. ,

460.12 Provide justification for concluding that the exhausts to the environment from the personnel areas in the Annex 1 building, electrical and mechanical equipment rooms in the Annex I and auxiliary buildings, and the diesel generator rooms will not be radioactive and, therefore, need not be monitored. (Sections 9.4 and 11.5) 460.13 Clarify whether the monitors provided in the exhaust ducts of the Annex 11 building, the fuel handling area of the auxiliary building, and the radiologically-controlled portion of the auxiliary building automatically facilitate connection of the applicable exhaust (i.e.,

monitor detects high radiation in the associated exhaust duct) to the containment air filtration system (Sections 9.4.3.1.2 and 11.5).

460.14 Clarify whether the steam generator blowdown system and component cooling water (CCW) radiation monitors provide any automatic control features. If not, indicate for what essential purpose these monitors are provided (e.g., manual actions to isolate the affected CCW loop or terminate SG blowdown on detection of high radiation by the subject monitor) (Section 11.5).

460,15 Section 9.3.3, 11.5.3, and 11.5.4 of the SSAR provide incomplete information on radiological sampling provisions for process and effluent streams. For example, the sampling provisions for the waste monitor tank contents, the detergent waste monitor tank contents, the steam generator blowdown, and the condenser air removal system have not been identified. Further, there is no reference to tritium measurements. Identify how the sample provisions for the liquid and gaseous process and effluent streams for the AP600 design meet the sampling provisions for such streams identified in Tables 1 and 2 of v

Section 11.5 of the SRP.

e

- . . = - . - -- - - .. . - - - . . . -

l i

1 460.16 Provide the following'information regarding accident monitoring instrumentation (Section 11.5). State if any of these items are outside the design scope of the AP600, but are withir. the design scope of the COL applicant.

i. . The recommended range for the noble gas effluent monitor for the l condenser ai removal system (Revision 3 to RG 1.97) is 10

pCi/cc to 10[ pCi/cc. The monitor is not needed if the effluent discharges through a common plant vent (however, this is not the case for the AP600). Table 11.5-1 of the SSAR provides a much narrower range, i.e.,10 pCi/cc to 10' pCi/cc. Why is the range so limited 7

b. Describe the calibration frequency and technique for calibrating the monitors.

c. Describe the methods used to ensure representative measurements are taken with appropriate background correction,

d. Describe the location of instrument readout (s) and the methods of recording this information, including the method or procedure for transmitting or disseminating the information or data,

e. Provide assurance of the capability to obtain readings at least every 15 minutes during and following an accident.

f. Describe the procedures or calculation methods to be used for converting instrument readings to release rates per unit time, based on exhaust air flow and consideration of radionuclide spectrum distribution as a function of time after shutdown.

g. Describe the sampling system design, including the sampling media, to demonstrate how the design meets the requirements identified in Clarification No. 2 of NUREG-0737, " Clarification of TMI Action Plan Requirements" (page ll.F.1-7).

h. Describe the sampling technique to be used under accident conditions to demonstrate how the technique meets the requirements identified in Clarification No. 3 of NUREG-0737 (pages II.F.1-7 and II.F.1-8).

i. Describe the sampling technique to ensure the system capability to collect and analyze or measure representative samples of radioactive iodines and particulates in plant gaseous effluents during and following an accident as identified-in Table II.F.1-2 of NUREG-0737 (page II.F.1-9).